10sep2014 Nuclear Fuel Cycle Notes Balakrishna Palanki

8/11/2019 10sep2014 Nuclear Fuel Cycle Notes Balakrishna Palanki

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The decay constant is given by = [ ln (N 0/N)] / t

or

= [ 2.303 log (N 0/N)] / t

Half life is the time required for the decay of half of the atoms originally present. t 1/2 = 0.693/

Human life can be measured as full life, since it is quite practical to measure. On the other hand, in an exponential function, it takesinfinite time to reach zero concentration as seen from the above table. Hence a practical measure has been the half life.

1.04 Mass defectIt has been found that the actual mass of an isotope of an element is less than the sum of masses of the protons, neutrons and electrons

present in it. This difference is called the mass defect. The mass defect is the loss of mass during the formation of the nucleus of theisotope.

1.05 Binding energyLoss of mass during the formation of the nucleus from nucleons is converted into energy. The release of the energy imparts stability tothe nucleus. The energy released when constituent nucleons combine to form a nucleus is called the binding energy of the nucleus. Inother words, energy equal to the binding energy will be needed to break up the nucleus into its constituent nucleons. The greater the

binding energy, the greater the stability of the nucleus. It is calculated by converting mass defect into energy, using Einst einsequation E = mc 2

1.06 Mean binding energy per nucleonThe binding energy of a nucleus divided by the number of the nucleons gives the mean binding energy per nucleon. In a graph ofmean binding energy per nucleon versus mass number, the energy first increases steeply, reaches a peak and then decreases. It isinteresting to note that the binding energies of intermediate elements (mass number 50 to 150) are higher than those of the heavierelements. For example, iron (mass number 56) has higher binding energy than uranium (mass number 235). An appreciable amount ofenergy will be released, when heavy nuclei such as uranium 235 (possessing lower binding energy) split up into nuclei of intermediateelements. This is the reason why heavy elements like U235 exhibit radio activity. Moreover, if such elements are split up by

bombardment with neutrons, a huge amount of energy is released. This is called nuclear fission process. Similarly, when lighterelements like deuterium are combined or fused to form heavier elements, a huge amount of energy is released. This is called nuclearfusion process.

Number ofhalf-liveselapsed

Fractionremaining

Percentageremaining

0 /1 1001 /2 502 253 /8 12 .54 /16 6 .255 /32 3 .1256 /64 1 .5637 /128 0 .781... ... ...n /2n 100/(2 n)


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1.07 Nuclear Fission versus Nuclear Fusion

Nuclear fission Nuclear fusion 1 Nuclear process where matter is converted to energy. Nuclear process where matter is converted to energy.2 It is a process of breaking a heavy nucleus with some properties

into two or more light fragments, with liberation of a largeamount of energy.

It is a process of fusing two light nuclei into a singlenucleus, with the liberation of a large amount of energy

3 Fission of U235 gives ~ 200 MeV One Deuterium -Tritium fusion gives ~ 17.6 MeV4 The process results in the emission of radioactive rays and in

long lived radio active waste.This process does not emit any kind of radioactive rays. Noradioactive waste. No melt down

5 This process takes place spontaneously at ordinary temperatures This process takes place at very high temperatures ( > 10

K)6 The mass number and the atomic number of the fission products(new elements) are considerably lower than that of the parentnucleus.

The mass number and the atomic number of the fusion products (new elements) are higher than that of the startingelements

7 It is a chain reaction It is not a chain reaction8 Fission also releases neutrons Fusion releases positrons9 The fission process is easy to control The fusion process is very difficult to control10 Resources limited, concentrated at specific locations Abundant and uniformly distributed11 Well developed and established technology Under development

1.08 Fissile and Fertile elements235U. 239Pu and 233U which are capable of undergoing fission upon bombardment by neutrons are called fissile elements.238U and 232Th which are capable of capturing a neutron without undergoing fission, are called fertile elements. The fertile elements238

U and232

Th, upon capture of a neutron undergo decay and transform to the fissile elements239

Pu and233

U respectively.

1.09 Approximate distribution of the energy released on fission of a heavy atom Kinetic Energy of fission product atoms ~ 169 MeVKinetic Energy of fission neutrons ~ 5 MeVInstantaneous gamma radiation ~ 5 MeVDelayed gamma radiation ~ 5 MeVFission products decay energy ~ 12 MeVBeta, gamma radiation after neutron capture ~ 8 MeV

1.10 Minimum Critical Masses of uranium at various U 235 enrichment levels

%U 235 Critical gross mass (U 235 + U 238 ), kg

For solution * For unreflected metal # For reflected metal#90 0.9 53 24.520 5.7 750 3755 38 3 114 1.8 708

* aqueous solutions, water reflected and optimum light water moderated # spherical shape


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2.4.1 ExplorationUranium is a slightly radioactive element that occurs throughout the Earth's crust. It is about 500 times more abundant than gold andabout as common as tin. It is present in most rocks and soils as well as in many rivers and in sea water. It is, for example, found inconcentrations of about four parts per million (ppm) in granite, which makes up 60% of the Earth's crust. In fertilisers, uraniumconcentration can be as high as 400 ppm (0.04%), and some coal deposits contain uranium at concentrations greater than 100 ppm(0.01%). Most of the radioactivity associated with uranium in nature is in fact due to other minerals derived from it by radioactivedecay processes, and which are left behind in mining and milling. There are a number of areas around the world where theconcentration of uranium in the ground is sufficiently high that extraction of it for use as nuclear fuel is economically feasible. Suchconcentrations are called ore.

Australia has the largest uranium resources in the world (31%), followed by Khazakstan (12%), Canada (9%), Russia, South Africa,Brazil, Namibia, USA (4%) and China. Uranium reserves are the amounts of ore that are estimated to be recoverable at stated costs.The low cost uranium reserves are expected to last over 80 years. The current annual world demand for uranium is 70000 tons.

A deposit of uranium, such as uraninite, discovered by geophysical techniques, is evaluated and sampled to determine the amounts ofuranium materials that are extractable at specified costs from the deposit. Uranium in nature consists primarily of two isotopes, U-238and U-235. The numbers refer to the atomic mass number for each isotope, or the number of protons and neutrons in the atomicnucleus. Naturally occurring uranium consists of approximately 99.28% U-238 and 0.71% U-235. The atomic nucleus of U-235 willnearly always fission when struck by a free neutron, and the isotope is therefore said to be a "fissile" isotope. The nucleus of a U-238atom on the other hand, rather than undergoing fission when struck by a free neutron, will nearly always absorb the neutron and yieldan atom of the isotope U-239. This isotope then undergoes natural radioactive decay to yield Pu-239, which, like U-235, is a fissileisotope. The atoms of U-238 are said to be fertile, because, through neutron irradiation in the core, some eventually yield atoms offissile Pu-239.

2.4.2 MiningBoth excavation and in situ techniques are used to recover uranium ore. Excavation may be underground and open pit

(surface) mining. In general, open pit mining is used where deposits are close to the surface and underground mining is used for deepdeposits, typically greater than 120 m deep. Open pit mines require large holes on the surface, larger than the size of the ore deposit,since the walls of the pit must be sloped to prevent collapse. As a result, the quantity of material that must be removed in order toaccess the ore may be large. Underground mines have relatively small surface disturbance and the quantity of material that must be
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removed to access the ore is considerably less than in the case of an open pit mine. Special precautions, consisting primarily ofincreased ventilation, are required in underground mines to protect against airborne radiation exposure.

An increasing proportion of the world's uranium (36%) now comes from in situ leach (ISL) mining, where oxygenatedgroundwater is circulated through a very porous ore body to dissolve the uranium oxide and bring it to the surface. ISL may be withslightly acid or with alkaline solutions to keep the uranium in solution. The uranium oxide is then recovered from the solution as in aconventional mill. ISL is of low cost and has minimal environmental impact.

The decision as to which mining method to use for a particular deposit is governed by the nature of the ore body, safety and economicconsiderations. An analysis of the depth of the overburden, the depth of the ore body, the grade of the ore, the inclination of the ore

body to the surface, and many other factors is required to decide which method (underground, open cut or in -situ leaching) is the most

economic.Uranium ore can be extracted through conventional mining in open pit and underground methods similar to those used for

mining other metals. In situ leach mining methods also are used to mine uranium in the United States. In this technology, uranium isleached from the in-place ore through an array of regularly spaced wells and is then recovered from the leach solution at a surface

plant. Uranium ores in the United States typically range from about 0.05 to 0.3% uranium oxide (U 3O8). Some uranium depositsdeveloped in other countries are of higher grade and are also larger than deposits mined in the United States. Uranium is also presentin very low-grade amounts (50 to 200 parts per million) in some domestic phosphate- bearing deposits of marine origin. Because verylarge quantities of phosphate-bearing rock are mined for the production of wet-process phosphoric acid used in high analysis fertilizersand other phosphate chemicals, at some phosphate processing plants the uranium, although present in very low concentrations, can beeconomically recovered from the process stream.

Uranium is mined from ores whose uranium content is on the order of 0.1 percent (one part per thousand). Most ore depositsare at or near the surface, and whether they are mined by open-pit or deep-mining techniques depends on the depth of the deposit andwhether it slopes downward. The ore is crushed and the uranium chemically extracted from it at the mouth of the mine.

2.4.3 MillingMined uranium ores normally are processed by grinding the ore materials to a uniform particle size and then treating the ore to extractthe uranium by chemical leaching. The milling process commonly yields dry powder-form material consisting of natural uranium,"yellowcake, " which is sold on the uranium market as U 3O8.

2.4.4 Uranium conversionMilled uranium oxide, U 3O8, must be converted to uranium hexafluoride, UF 6, which is the form required by most commercialuranium enrichment facilities currently in use. A solid at room temperature, uranium hexafluoride can be changed to a gaseous form atmoderately higher temperature of 57 C. The uranium hexafluoride conversion product contains only natural, not enriched, uranium.Triuranium octaoxide (U 3O8) is also converted directly to ceramic grade uranium dioxide (UO 2) for use in reactors not requiringenriched fuel, such as CANDU. The volumes of material converted directly to UO 2 are typically quite small compared to the amountsconverted to UF 6.
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2.4.5 EnrichmentThe concentration of the fissionable isotope, U-235 (0.71% in natural uranium) is less than that required to sustain a nuclear chainreaction in light water reactor cores. Natural UF 6 thus must be enriched in the fissionable isotope for it to be used as nuclear fuel. Thedifferent levels of enrichment required for a particular nuclear fuel application are specified by the customer: light-water reactor fuelnormally is enriched to 3.5% U-235, but uranium enriched to lower concentrations also is required. Enrichment is accomplished usingsome one or more methods of isotope separation.

Isotope separation is a difficult and energy intensive activity. Enriching uranium is difficult because the two isotopes have very nearlyidentical chemical properties, and are very similar in weight: 235U is only 1.26% lighter than 238U. Enrichment methods exploit theslight differences in atomic weights of the various isotopes. A feature common to all large-scale enrichment schemes is that theyemploy a number of identical stages which produce successively higher concentrations of 235 U. Each stage concentrates the product ofthe previous step further before being sent to the next stage. Similarly, the tailings from each stage are returned to the previous stagefor further processing. This sequential enriching system is called a cascade.

There are currently two generic commercial methods employed internationally for enrichment: gaseous diffusion (referred to as first generation) and gas centrifuge ( second generation). In gaseous diffusion, natural uranium in the form of uranium hexafluoride gas(UF 6), a product of chemical conversion, is forced under pressure to migrate through a porous barrier. The molecules of

235UF 6 penetrate the barrier slightly faster than those of 238UF 6. The process has to be repeated several times.

The old GD plants are being replaced by new GC plants which are more energy efficient and can be built in modules as demandincreases. For example, the energy use for one SWU in a GC plant is only 50 KWh compared with 2500 KWh in an old GD plant.

Laser generation methods will become established because they are more efficient in terms of the energy input for the same degree ofenrichment and the next method of enrichment to be commercialized will be referred to as third generation.

Gaseous diffusion (energy intensive) and gas centrifuge are the commonly used uranium enrichment technologies, but new enrichmenttechnologies are currently being developed: Laser enrichment(South Africa), chemical exchange (Japanese and French),aerodynamic (UF 6 + H 2 carrier gas).
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The bulk (96%) of the byproduct from enrichment is depleted uranium (DU), which can be used for armor, kinetic energy penetrators,radiation shielding and ballast. Still, there are vast quantities of depleted uranium in storage. The United States Department of Energyalone has 470,000 tonnes. About 95% of depleted uranium is stored as uranium hexafluoride (UF 6).

2.4.6 FabricationFor use as nuclear fuel, enriched uranium hexafluoride is converted into uranium dioxide (UO 2) powder that is then processed into

pellet form. The pellets are then fired in a high temperature sintering furnace to create hard, ceramic pellets of enriched uranium. Thecylindrical pellets then undergo a grinding process to achieve a uniform pellet size. The pellets are stacked, according to each nuclearreactor core' s design specifications, into tubes of corrosion-resistant metal alloy. The tubes are sealed to contain the fuel pellets: thesetubes are called fuel rods. The finished fuel rods are grouped in special fuel assemblies that are then used to build up the nuclear fuelcore of a power reactor.

The metal used for the tubes depends on the design of the reactor. Stainless steel was used in the past, but most reactors now usezirconium. For the most common types of reactors, boiling water reactors (BWR) and pressurized water reactors (PWR), the tubes areassembled into bundle s[9] with the tubes spaced precise distances apart. These bundles are then given a unique identification number,which enables them to be tracked from manufacture through use and into disposal.

2.4.7 Transport of radioactive materialsTransport is an integral part of the nuclear fuel cycle. There are nuclear power reactors in operation in several countries but uraniummining is viable in only a few areas. Also, in the course of over forty years of operation by the nuclear industry, a number ofspecialized facilities have been developed in various locations around the world to provide fuel cycle services and there is a need totransport nuclear materials to and from these facilities. Most transports of nuclear fuel material occur between different stages of thecycle, but occasionally a material may be transported between similar facilities. With some exceptions, nuclear fuel cycle materials aretransported in solid form, the exception being uranium hexafluoride (UF 6) which is considered a gas. Most of the material used innuclear fuel is transported several times during the cycle. Transports are frequently international, and are often over large distances.

Nuclear materials are generally transported by specialized transport companies.Since nuclear materials are radioactive, it is important to ensure that radiation exposure of both those involved in the transport of suchmaterials and the general public along transport routes is limited. Packaging for nuclear materials includes, where appropriate,shielding to reduce potential radiation exposures. In the case of some materials, such as fresh uranium fuel assemblies, the radiationlevels are negligible and no shielding is required. Other materials, such as spent fuel and high-level waste, are highly radioactive andrequire special handling. To limit the risk in transporting highly radioactive materials, containers known as spent nuclear fuel shippingcasks are used which are designed to maintain integrity under normal transportation conditions and during hypothetical accidentconditions.

2.4.8 In-core fuel managementFuel is loaded into a reactor in a careful pattern so as to obtain the most energy production from it before it becomes no longer usable.

Fresh fuel is more reactive than old fuel, and this reactivity is used to keep the reactor critical. Typically, a reactor is fueled in cycles,each cycle lasting one to two years, and a fuel batch is kept in the reactor for three or four cycles.

A nuclear reactor core is composed of a few hundred "assemblies", arranged in a regular array of cells, each cell being formed by afuel or control rod surrounded, in most designs, by a moderator and coolant, which is water in most reactors.

Because of the fission process that consumes the fuels, the old fuel rods must be changed periodically to fresh ones (this period iscalled a cycle). However, only a part of the assemblies (typically one-third) are removed since the fuel depletion is not spatiallyuniform. Furthermore, it is not a good policy, for efficiency reasons, to put the new assemblies exactly at the location of the removedones. Even bundles of the same age may have different burn-up levels, which depends on their previous positions in the core. Thus theavailable bundles must be arranged in such a way that the yield is maximized, while safety limitations and operational constraints aresatisfied. Consequently reactor operators are faced with the so-called optimal fuel reloading problem , which consists in optimizingthe rearrangement of all the assemblies, the old and fresh ones, while still maximizing the reactivity of the reactor core so as tomaximise fuel burn-up and minimise fuel-cycle costs.This is a discrete optimization problem, and computationally infeasible by current combinatorial methods, due to the huge number of

permutations and the complexity of each computation. Many numerical methods have been proposed for solving it and manycommercial software packages have been written to support fuel management. This is an on-going issue in reactor operations as nodefinitive solution to this problem has been found and operators use a combination of computational and empirical techniques tomanage this problem.

2.4.9 Unloading and coolingSpent reactor fuel is extremely radioactive, and its radioactivity also makes it a source of heat. When the spent fuel is removed fromthe reactor, it must continue to be both shielded and cooled. This is accomplished by placing the spent fuel in a water storage poollocated next to the reactor. The water in the pool contains a large amount of dissolved boric acid .
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2.4.10 ReprocessingBoth the converted plutonium and residual uranium-235 in spent fuel can be recycled. Such materials can be recovered by chemicallyreprocessing the fuel. Equally as significant, reprocessing can reduce the volume and radioactivity of the waste material, which mustultimately be eliminated by some method of permanent disposal.

Spent fuel discharged from reactors contains appreciable quantities of fissile (U-235 and Pu-239), fertile (U-238), and otherradioactive materials, including reaction poisons, which is why the fuel had to be removed. These fissile and fertile materials can bechemically separated and recovered from the spent fuel. The recovered uranium and plutonium can, if economic and institutionalconditions permit, be recycled for use as nuclear fuel. This is currently not done for civilian spent nuclear fuel in the US. Mixed oxide, or MOX fuel, is a blend of reprocessed uranium and plutonium and depleted uranium which behaves similarly, althoughnot identically, to the enriched uranium feed for which most nuclear reactors were designed. MOX fuel is an alternative to low-enriched uranium (LEU) fuel used in the light water reactors which predominate nuclear power generation.Currently, plants in Europe are reprocessing spent fuel from utilities in Europe and Japan. Reprocessing of spent commercial-reactornuclear fuel is currently not permitted in the United States due to the perceived danger of nuclear proliferation. However the recentlyannounced Global Nuclear Energy Partnership would see the U.S. form an international partnership to see spent nuclear fuelreprocessed in a way that renders the plutonium in it usable for nuclear fuel but not for nuclear weapons.

2.4.11 Partitioning and transmutationAs an alternative to the disposal of the PUREX raffinate in glass or Synroc, the most radiotoxic elements can be removed throughadvanced reprocessing. After separation, the minor actinides and some long lived fission products can be converted to short-livedisotopes by either neutron or photon irradiation. This is called transmutation.

2.4.12 Waste conditioning and disposalIn the absence of reprocessing, the spent fuel is considered to be waste and must be prepared for disposal. This operation is to be

performed in a separate facility, for which the Department of Energy has responsibility in the United States. As of 1998, thedepartment is to begin receiving spent fuel from utilities largely on an oldest -fuel- first schedule.

When a holistic view is taken of the nuclear waste disposal process, the risks seem extremely small, yet among the general publicthese risks are one of the most feared aspects of the nuclear fuel cycle. A great deal of suspicion about the process arises from thenumerous incidents of mismanagement of other types of waste, and these fears have been encouraged by antinuclear activists.

The waste disposal method currently being planned by all countries with nuclear power plants is called geologic disposal. This meansthat all conditioned nuclear wastes are to be deposited in mined cavities deep underground. Shafts are to be sunk into a solid rockstratum, with tunnel corridors extending horizontally from the central shaft region and tunnel.

A current concern in the nuclear power field is the safe disposal and isolation of either spent fuel from reactors or, if the reprocessing

option is used, wastes from reprocessing plants. These materials must be isolated from the biosphere until the radioactivity containedin them has diminished to a safe level. In the U.S., under the Nuclear Waste Policy Act of 1982 as amended, the Department ofEnergy has responsibility for the development of the waste disposal system for spent nuclear fuel and high-level radioactive waste.Current plans call for the ultimate disposal of the wastes in solid form in a licensed deep, stable geologic structure called a deepgeological repository. The Department of Energy chose Yucca Mountain as the location for the repository. However, its opening has

been repeatedly delayed.It is worth noting that some non -PLWR reactor designs, and in particular the ones using liquid thorium fuel in molten salt reactors,would produce virtually no long-lasting nuclear waste. It is also possible to burn rather than bury nuclear waste for instance in IntegralFast Reactor or in variation of molten salt reactor.

2.4.13 Uranium from other sourcesSea water contains uranium at 3.5 ppb, which means that 4.5 billion tons of uranium are available in sea water. This is many timesmore than all the terrestrial resources of uranium. ORNL claim to have developed high capacity reusable adsorbents and high surfacearea polyethylene fibers that have lowered the high cost of uranium extraction from sea water. The adsorbents are made by subjectinghigh-surface area polyethylene fibers to ionizing radiation, then reacting these pre-irradiated fibers with chemical compounds thathave a high affinity for selected metals.Uranium may also be obtained as a byproduct in the extraction of copper and gold and from phosphoric acid.
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3 Nuclear Reactors Encyclopedia Britannica

3.1 Nuclear ReactorsAny of a class of devices that can initiate and control a self-sustaining series of nuclear fissions is a nuclear reactor. Such devices areused as research tools, as systems for producing radioisotopes, and most prominently as energy sources. The latter are commonlycalled power reactors.

3.2 Principles of operation Nuclear reactors operate on the principle of nuclear fission, the process in which a heavy atomic nucleus splits into two smaller

fragments. The nuclear fragments are in very excited states and emit neutrons , other subatomic particles , and photons . The emittedneutrons may then cause new fissions, which in turn yield more neutrons, and so forth. Such a continuous self-sustaining series offissions constitutes a fission chain reaction. A large amount of energy is released in this process, and this energy is the basis of nuclear

power systems.

In an atomic bomb the chain reaction is designed to increase in intensity until much of the material has fissioned. This increase is veryrapid and produces the extremely prompt, tremendously energetic explosions characteristic of such bombs. In a nuclear reactor thechain reaction is maintained at a controlled, nearly constant level. Nuclear reactors are so designed that they cannot explode likeatomic bombs.

Most of the energy of fission approximately 85 percent of it is released within a very short time after the process has occurred. Theremainder of the energy produced as a result of a fission event comes from the radioactive decay of fission products, which are fissionfragments after they have emitted neutrons. Radioactive decay is the process by which an atom reaches a more stable state; the decay

process continues even after fissioning has ceased, and its energy must be dealt with in any proper reactor design.

3.2.1 Chain reaction and criticalityThe course of a chain reaction is determined by the probability that a neutron released in fission will cause a subsequent fission. If theneutron population in a reactor decreases over a given period of time, the rate of fission will decrease and ultimately drop to zero. Inthis case the reactor will be in what is known as a subcritical state. If over the course of time the neutron population is sustained at aconstant rate, the fission rate will remain steady, and the reactor will be in what is called a critical state. Finally, if the neutron

population increases over time, the fission rate and power will increase, and the reactor will be in a supercritical state.

Before a reactor is started up, the neutron population is near zero. During reactor start-up, operators remove control rods from the corein order to promote fissioning in the reactor core, effectively putting the reactor temporarily into a supercritical state. When the reactorapproaches its nominal power level, the operators partially reinsert the control rods, balancing out the neutron population over time. Atthis point the reactor is maintained in a critical state, or what is known as steady-state operation. When a reactor is to be shut down,operators fully insert the control rods, inhibiting fission from occurring and forcing the reactor to go into a subcritical state.

3.2.2 Reactor controlA commonly used parameter in the nuclear industry is reactivity, which is a measure of the state of a reactor in relation to where itwould be if it were in a critical state. Reactivity is positive when a reactor is supercritical, zero at criticality, and negative when thereactor is subcritical. Reactivity may be controlled in various ways: by adding or removing fuel, by altering the ratio of neutrons thatleak out of the system to those that are kept in the system, or by changing the amount of absorber that competes with the fuel forneutrons. In the latter method the neutron population in the reactor is controlled by varying the absorbers, which are commonly in theform of movable control rods (though in a less commonly used design, operators can change the concentration of absorber in thereactor coolant). Changes of neutron leakage, on the other hand, are often automatic. For example, an increase of power will cause areactors coolant to reduce in density and possibly boil. This decrease in coolant density will increase neutron leakage out of thesystem and thus reduce reactivity a process known as negative-reactivity feedback. Neutron leakage and other mechanisms ofnegative-reactivity feedback are vital aspects of safe reactor design.

A typical fission interaction takes place on the order of one picosecond (10 12 second). This extremely fast rate does not allow enoughtime for a reactor operator to observe the systems state and respond appropriately. Fortunately, reactor control is aided by the

presence of so-called delayed neutrons, which are neutrons emitted by fission products some time after fission has occurred. Theconcentration of delayed neutrons at any one time (more commonly referred to as the effective delayed neutron fraction) is less than 1

percent of all neutrons in the reactor. However, even this small percentage is sufficient to facilitate the monitoring and control ofchanges in the system and to regulate an operating reactor safely.

3.2.3 Fissile and fertile materialsAll heavy nuclides can fission if they are in an excited enough state, but only a few fission readily and consistently when struck byslow (low-energy) neutrons. Such species of atoms are called fissile. The most important of these are uranium-233 ( 233U), uranium-235 (235U), plutonium-239 ( 239Pu), and plutonium-241 ( 241Pu). The only one that occurs in usable amounts in nature is uranium-235,which makes up a mere 0.711 percent of natural uranium.
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3.2.4 Principles of operation3.2.4.1 Heat removalThe energy of fission is quickly converted to heat, the bulk of which is deposited in the fuel. A coolant is therefore required to removethis heat. The most common coolant is water , but any fluid can be used. Heavy water (deuterium oxide), air, carbon dioxide, helium,liquid sodium, sodium-potassium alloy (called NaK), molten salts, and hydrocarbons have all been used in reactors.

Some research reactors are operated at very low power and have no need for a dedicated cooling system; in such units the smallamount of generated heat is removed by conduction and convection to the environment. Very high power reactors, on the other hand,must have extremely sophisticated cooling systems to remove heat quickly and reliably; otherwise, the heat will build up in the reactorfuel and melt it. Indeed, most reactors operate on the principle that their fuel cannot be allowed to melt; therefore, the systemsdesigned to cool the fuel must operate sufficiently under both normal and abnormal conditions. Systems that enable sufficient coolingduring all credible abnormal conditions in nuclear power plants are referred to as emergency core-cooling systems.

3.2.4.2 ShieldingAn operating reactor is a powerful source of radiation, since fission and subsequent radioactive decay produce neutrons and gammarays, both of which are highly penetrating radiations. A reactor must have specifically designed shielding around it to absorb andreflect this radiation in order to protect technicians and other reactor personnel from exposure. In a popular class of research reactorsknown as swimming pools, this shielding is provided by placing the reactor in a large, deep pool of water. In other kinds o f reactors,the shield consists of a thick concrete structure around the reactor system referred to as the biological shield. The shield also maycontain heavy metals, such as lead or steel, for more effective absorption of gamma rays, and heavy aggregates may be used in theconcrete itself for the same purpose.

3.2.4.3 Critical concentration and size Not every arrangement of material containing fissile fuel can be brought to criticality. Even if a reactor was designed such that noneutrons could leak out, a critical concentration of fissile material would have to be present in order to bring the reactor to a criticalstate. Otherwise, absorption of neutrons by other constituents of the reactor might dominate and inhibit a sustained chain fissionreaction. Similarly, even where there is a high-enough concentration for criticality, the reactor must occupy an appropriate volume and

be of a prescribed geometric form, or else more neutrons will leak out than are created through fission. This requirement imposes alimit on the minimum critical volume and critical mass within a reactor.

Although the only useful fissile material in nature, uranium 235, is found in natural uranium, there are only a few combinations andarrangements of this and other materials that enable a reactor to maintain a critical state for a period of time. To increase the range offeasible reactor designs, enriched uranium is often used. Most of todays power reactors employ enriched uranium fuel in which the

percentage of uranium-235 has been increased to between 3 and 5 percent, approximately five and a half times the concentration innatural uranium.

3.2.5 Thermal, intermediate, and fast reactorsReactors are conveniently classified according to the typical energies of the neutrons that cause fission. Neutrons emanating in fissionare very energetic; their average energy is around two million electron volts (2 MeV), 80 million times higher than the energy ofatoms in ordinary matter at room temperature. As the neutrons scatter or collide with nuclei in a reactor, they lose energy. This actionis referred to as down-scattering; The choice of reactor materials and of fissile material concentrations determines the rate at whichneutrons are slowed through down-scattering before causing fission.

3.3 Reactor design and componentsThere are a large number of ways in which a reactor may be designed and constructed, and many types have been experimentallyrealized. Over the years, nuclear engineers have developed reactors with solid fuels and liquid fuels, thick reflectors and no reflectors,forced cooling circuits and natural conduction or convection heat-removal systems, and so on.

3.3.1 CoreAll reactors have a core, a central region that contains the fuel, fuel cladding, coolant, and, where separate from the latter, moderator.It is in the core that fission occurs and the resulting neutrons migrate.

The fuel is usually heterogeneous i.e., it consists of elements containing fissile material along with a diluent. This diluting agent may be fertile material or simply material that has good mechanical and chemical properties and does not readily absorb neutrons. Alldiluents act as a matrix in which the fissile material can stably reside through its operable life. In solid fuels, the diluted fissile materialis enclosed in a cladding a substance that isolates the fuel from the coolant and minimizes the likelihood that radioactive fission

products will be released. Cladding is often referred to as a reactors first fission product barrier, as it is the first barrier that fissilematerial contacts after nuclear fission.
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3.3.1.1 Fuel typesThe light water reactor (LWR), which is the most widely used variety for commercial power generation in the world, employs a fuelconsisting of pellets of sintered uranium dioxide loaded into cladding tubes of zirconium alloy or some other advanced claddingmaterial. The tubes, called pins or rods, measure approximately 1 cm (less than half an inch) in diameter and roughly 3 to 4 metres (10to 13 feet) in length. The tubes are bundled together into a fuel assembly, with the pins arranged in a square lattice. The uranium usedin the fuel is 3 to 5 percent enriched. Since light (ordinary) water, used in LWRs as both the coolant and the moderator, tends toabsorb more neutrons than other moderators do, such enrichment is crucial.

The CANDU which is the principal type of heavy water, uses natural uranium compacted into pellets. These pellets are inserted inlong tubes and arranged in a lattice. A CANDU reactor fuel assembly measures approximately 1 metre (almost 40 inches) in length.

Several assemblies are arranged end-to-end within a channel inside the reactor core. The use of heavy water rather than light water asthe moderator enhances the scattering of neutrons rather than their capture, thereby increasing the probability of fission with the fuelmaterial.

In one version of the high-temperature graphite reactor, the fuel is constructed of small spherical particles, or microspheres, containinguranium dioxide at the centre with concentric shells of carbon, silicon carbide, and carbon around them. These shells serve aslocalized cladding for each fuel sphere. The particles are then mixed with graphite and encased in a macroscopic graphite cladding.

In a sodium cooled fast reactor, commonly called a liquid-metal reactor (LMR), the fuel consists of uranium dioxide or uranium- plutonium dioxide pellets (French design) or of uranium-plutonium -zirconium metal alloy pins (U.S. design) in steel cladding.

The most common type of fuel used in research reactors consists of plates of a uranium-aluminium alloy with an aluminum cladding.The uranium is enriched to slightly less than 20 percent, while silicon and aluminum are in cluded in the meat of the plate to serve asthe diluent and fuel matrix. Although aluminum has a lower melting point than other cladding materials, the flat plate design maintainsa low fuel temperature, as the plates are often barely more than 1.25 mm thick. A common variety of research reactors known as

TRIGA (from training, research, and isotope-production reactors General Atomic) employs a fuel of mixed uranium and zirconiumhydride, often doped with small concentrations of erbium and the whole clad in stainless steel.

3.3.1.2 Coolants and moderatorsA variety of substances, including light water, heavy water, air, carbon dioxide, helium, liquid sodium, liquid sodium-potassium alloy,and hydrocarbons (oils), have been used as coolants. Such substances are good conductors of heat and serve to carry the thermalenergy produced by fission from the core to the steam-generating equipment of the nuclear power plant. In many cases, the samesubstance functions as the coolant as well as moderator as in the case of light and heavy water.

3.3.2 ReflectorA reflector is a region of unfueled material surrounding the core. Its function is to scatter neutrons that leak from the core and therebyreturn some of them to the core. This reduces core size and smooths out the power density. The reflector is particularly important inresearch reactors, since it is the region in which much of the experimental apparatus is located.

In most types of power reactors, a reflector is less important; this is due to the reactors large size, which reduces the proportion ofneutrons that may leak from the core region. The liquid-metal reactor represents a special case. Most sodium-cooled reactors aredeliberately built to allow a large fraction of their neutrons those not needed to maintain the chain reaction to leak from the core.These neutrons are valuable because they can produce new fissile material if they are absorbed by fertile material. Thus, fertilematerial generally depleted uranium or its dioxide is placed around the core to catch the leaking neutrons. Such an absorbingreflector is referred to as a blanket or a breeding blanket.

3.3.3 Reactor control elementsAll reactors need special elements for control. Although control can be achieved by varying parameters of the coolant circuit or byvarying the amount of absorber dissolved in the coolant or moderator, by far the most common method involves the use of specialabsorbing assemblies namely, control rods or sometimes blades. Typically a reactor is equipped with three types of rods for different

purposes: (1) safety rods for starting up and shutting down the reactor, (2) regulating rods for adjusting the reactors powe r rate, and(3) shim rods for compensating for changes in reactivity as fuel is depleted by fission and neutron capture.

3.3.4 Structural componentsThese are the parts of a reactor system that hold the reactor together and permit it to function as a useful energy source. The mostimportant structural component is usually the reactor vessel. In both the light-water reactor and the high-temperature gas-controlledreactor (HTGR), a pressure vessel is used so that the coolant can be contained and operated under conditions appropriate for powergeneration namely, elevated temperature and pressure. Within the reactor vessel are a number of structural elements: grids forholding the reactor core and solid reflectors, control-rod guide tubes, internal thermal hydraulic components (e.g., pumps or steamcirculators) in some cases, instrument tubes, and components of safety systems.


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3.3.5 Coolant systemThe function of a power reactor installation is to extract as much heat of nuclear fission as possible and convert it to useful power,generally electricity. The coolant system plays a pivotal role in performing this function. A coolant fluid enters the core at lowtemperature and exits at a higher temperature after collecting the fission energy. This higher-temperature fluid is then directed toconventional thermodynamic components where the heat is converted into electric power. In most light-water, heavy-water, and gas-cooled power reactors, the coolant is maintained at high pressure. Sodium and organic coolants operate at atmospheric pressure.

3.3.6 Containment systemReactors are designed with the expectation that they will operate safely without releasing radioactivity to their surroundings. It is,

however, recognized that accidents can occur. An approach using multiple barriers has been adopted to deal with such accidents.These barriers are, successively, the fuel cladding, primary vessel, and thick shielding.

3.4 Types of reactorsMost of the world's existing reactors are power reactors. There also are many research reactors, and the navies of many nations includesubmarines and surface ships driven by propulsion reactors. There are several types of power reactors, but only one, the light-waterreactor, is widely used. Accordingly, this variety is discussed in considerable detail here.

3.4.1 Power reactors3.4.1.1 Light-water reactorLWRs are power reactors that are cooled and moderated with ordinary water. There are two basic types: the pressurized-water reactor(PWR) and the boiling-water reactor (BWR). In the first type, high-pressure, high-temperature water removes heat from the core andis then passed to a steam generator. There the heat from the primary loop is transferred to a lower-pressure secondary loop alsocontaining water. The water in the secondary loop enters the steam generator at a pressure and temperature slightly below thatrequired to initiate boiling. Upon absorbing heat from the primary loop, however, it becomes saturated and ultimately slightlysuperheated. The steam thus generated ultimately serves as the working fluid in a steam-turbine cycle.

A BWR operates on the principle of a direct power cycle. Water passing through the core is allowed to boil at an intermediate pressurelevel; the saturated steam that exits the core region is transported through a series of separaters and driers located within the reactorvessel that promote a superheated state. The superheated water vapour is then used as the working fluid to turn the steam turbine.
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Advantages and disadvantagesEach LWR design has its own advantages and disadvantages, and as a result, a competitive economic market has existed between theBWR and PWR concepts since the 1960s. For instance, although there are fewer mechanical components in the steam cycle of a BWRdesign, additional components are required to support the reactors emergency core -cooling system. Furthermore, the BWR vesselsinternal system is more complex, since it includes internal recirculation pumps and complex steam separation and drying equipmentthat are not found in a PWR design. On the other hand, even though the internals of the PWR are simpler, a BWR power plant issmaller, because it has no steam generators. In fact, the steam generators of a PWR there are typically four of them in a big plant are larger than the reactor vessel itself.

The direct-cycle philosophy of a BWR design reduces heat loss between the core and the steam turbine, but the BWR operates atlower pressures and temperatures than the PWR, giving it less thermodynamic efficiency. Furthermore, because the BWRs powerdensity is somewhat lower than that of the PWR, the pressure vessel must be built to a larger diameter for the same reactor power. Onthe other hand, because the BWR operates at lower pressure, its pressure vessel is thinner than the pressure vessel of a PWR.

3.4.1.2 High-temperature gas-cooled reactorThe HTGR is fueled with a mixture of graphite and fuel-bearing microspheres. There are two competitive designs of this reactor type:(1) a German system that uses spherical fuel elements of tennis-ball size loaded into a graphite silo and (2) an American version inwhich the fuel is loaded into precisely located graphite hexagonal prisms.

3.4.1.3 Liquid-metal reactorsSodium-cooled, fast-neutron-spectrum reactors received much attention during the 1960s and '70s when it appeared that their breedingcapabilities would soon be needed to supply fissile material to a rapidly expanding nuclear industry. When it became clear in the1980s that this was not a realistic expectation, enthusiasm slackened.

3.4.1.4 CANDU reactorCanada focused its developmental efforts on reactors that would utilize abundant domestic natural uranium as fuel without having toresort to enrichment services that could be supplied only by other countries. The result of this policy was CANDU the line of naturaluranium-fueled reactors moderated and cooled by heavy water .

3.4.1.5 Advanced gas-cooled reactorThe advanced gas-cooled reactor (AGR) was developed in Britain as the successor to reactors of the Calder Hall class, whichcombined plutonium production and power generation. Calder Hall was the first nuclear station to feed an appreciable amount of

power into a civilian network. It was fueled with slugs of natural uranium metal canned in aluminum, cooled with carbon dioxide andemployed a moderator consisting of a block of graphite pierced by fuel channels. In the AGR, fuel pins clad in Zircaloy (a trademarkfor alloys of zirconium having low percentages of chromium, nickel, iron, and tin) and loaded with approximately 2 percent enricheduranium dioxide are placed into zirconium-alloy channels that pierce a graphite moderator block. This design utilizes fast neutronenergies and is therefore referred to as a fast reactor. The enriched fuel permits operation to economic levels of fuel burnup. A coolant
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of carbon dioxide absorbs and transports heat to a steam generator, where the heat is conducted to the secondary loop and ultimately toa traditional steam-turbine cycle. Although a number of AGRs have been built in the United Kingdom, maintenance and malfunctionissues have proved to be more costly than expected, and no new AGRs are projected for construction

3.4.2 Research reactors

3.4.2.1 Water-cooled, plate-fuel reactorThis is the most common type of research reactor. It uses enriched uranium fuel in plate assemblies and is cooled with water. Water-cooled, plate-fuel reactors operate over a wide range of thermal power levels, from a few kilowatts to hundreds of megawatts. Thesystems with the lowest power ratings are usually operated at universities and used primarily for teaching, whereas those with thehighest are used by research laboratories chiefly for materials testing and characterization as well as for general research.

A common form of the water-cooled, plate-fuel reactor is the pool reactor, in which the reactor core is positioned near the bottom of alarge, deep pool of water. This has the advantage of simplifying both observation and the placement of channels, commonly referredto as beam ports, from which beams of neutrons can be directed and transported. At lower thermal power levels, no pumping isrequired, as the natural convection of the coolant past the fuel plates provides sufficient heat removal to maintain a safe operatingstate. A heat exchanger is usually located at or near the top of the pool, where the hottest water is stratified. At higher operating powerlevels, pumping becomes necessary to augment the natural circulation .

3.4.2.2 TRIGA reactorsThe training, research, and isotope-production reactors General Atomic (TRIGA) system is a popular variety of research reactor. It isanother tank-type water-cooled system, but its fuel differs from that employed by the plate-fuel research reactors described above. The

fuel element of the TRIGA reactor consists of stainless steel or aluminium clad rods containing mixed uranium andzirconium hydrides that are often doped with small concentrations of erbium. In contrast to thin plate-type fuel, TRIGA fuel elementsare nominally 3.8 cm (1.5 inches) in diameter and in general approximately 67 cm (26 inches) in total length. A unique characteristicof this fuel is that it exhibits an extremely large negative power-reactivity coefficient so large that the TRIGA reactor can be placedin an extremely supercritical state for an instant, causing its power to rise very rapidly, after which it quickly shuts itself down on the

basis of the fuels inherent material composition and characteristics. The resulting power transient is referred to as a pulse and isuseful for a number of dynamic experiments that require large bursts of neutrons over a short period of time. The total energy releasedin a pulse does not draw concern toward the safety of the reactor, since the automatic shutdown occurs very quickly and the energyrelease is proportional to both peak power and pulse duration.

3.4.2.3 Other research reactorsAs in the case of power reactors, a number of different reactor types have seen service as research reactors, and some are still inoperation. The variety is so great as to defy cataloging. There have been homogeneous (fueled solution cores), fast, graphite-

moderated, heavy-water-moderated, and beryllium-moderated reactors, as well as those adapted to use fuels left over from powerreactor experiments.

3.4.3 Ship propulsion reactorsThe original, and still the major, naval application of nuclear energy is the propulsion of submarines . The chief advantage of usingnuclear reactors for submarine propulsion is that they, unlike fossil-fuel combustion systems, require no air for power generation.Consequently, a nuclear-powered submarine can remain underwater for prolonged periods, whereas a conventional submarine has toresurface for air needed for combustion.

3.4.4 Production reactorsThe very first nuclear reactors were built for the express purpose of manufacturing plutonium for nuclear weapons , and theeuphemism of calling them production reactors has persisted to this day. At present, most of the material produced by such systems istritium (3H, or T), the fuel for hydrogen bombs.

3.4.5 Specialized reactors Nuclear reactors have been developed to provide electric power and steam heat in far-removed, isolated areas. Russia , for instance,operates smaller power reactors specially designed to supply both electricity and steam for heating to accommodate the needs of anumber of remote Arctic communities.

3.5 Reactor safety Nuclear reactors contain very large amounts of radioactive isotopes mostly fission products but also such heavy elements as plutonium. If this radioactivity were to escape the reactor, its effects on the people in the vicinity would be severe. The deleteriouseffects of exposure to high levels of ionizing radiation would include increased rates of cancer.
http://www.britannica.com/eb/article-9110183http://www.britannica.com/eb/article-9110183http://www.britannica.com/eb/article-9110183http://www.britannica.com/eb/article-9060467http://www.britannica.com/eb/article-9060467http://www.britannica.com/eb/article-9060467http://www.britannica.com/eb/article-9110178http://www.britannica.com/eb/article-9110178http://www.britannica.com/eb/article-9110178http://www.britannica.com/eb/article-9073439http://www.britannica.com/eb/article-9073439http://www.britannica.com/eb/article-9105999http://www.britannica.com/eb/article-9105999http://www.britannica.com/eb/article-9105999http://www.britannica.com/eb/article-9105999http://www.britannica.com/eb/article-9073439http://www.britannica.com/eb/article-9110178http://www.britannica.com/eb/article-9060467http://www.britannica.com/eb/article-9110183


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3.5.1 Preventive measuresSince no human activity can be shown to be absolutely safe, all these measures cannot reduce the risks to zero, but it is the aim of therules and safety systems to minimize the risk to the point where a reasonable individual would conclude they are trivial.

3.5.2 Mitigating measuresTwo of the principal safety measures, the safety rods and the containment structure, have already been described. Other major safetysystems are the emergency core cooling system, which makes it possible to cool the reactor if normal cooling is disrupted, and theemergency power system, which is designed to supply electrical power in case the normal supply is disrupted. ( Material from

Encyclopedia ends here)

3.5.3 Some useful parameters of reactor performance

Online hours: The total clock hours in the reporting period, during which the Unit operated with the generator breaker closed to thestation bus.Offline hours: The difference between the total clock hours in the reporting period and the online hours.Availability factor: It is the ratio between the number of hours the Unit was online and the total number of hours in the reporting

period.Capacity factor: ratio of the energy actually produced during the reporting period to the energy that could have been produced atmaximum capacity under continuous operation during the whole of the reporting period.Number of outages: The number of times the generator breaker opened when the unit was feeding electrical energy to the grid.Full power days: Fission power of reactor multiplied by time in hours during which reactor was operating at this power divided by(100% fission power x 24)

p i ti/ (P max ti)3.5.4 Operating Indian Power Reactors

S.No Plant Unit Type MWe Date of commercial operation1 Tarapore APS 1 BWR 160 28.10.19692 Tarapore APS 2 BWR 160 28.10.19693 Tarapore APS 3 PHWR 540 18.08.20064 Tarapore APS 4 PHWR 540 12.09.20055 Rajasthan APS 1 PHWR 100 16.12.19736 Rajasthan APS 2 PHWR 200 01.04.19817 Rajasthan APS 3 PHWR 220 01.06.20008 Rajasthan APS 4 PHWR 220 23.12.2000

9 Rajasthan APS 5 PHWR 220 04.02.201010 Rajasthan APS 6 PHWR 220 31.03.201011 Madras APS 1 PHWR 220 27.01.198412 Madras APS 2 PHWR 220 21.03.198613 Kaiga GS 1 PHWR 220 16.11.200014 Kaiga GS 2 PHWR 220 16.03.200015 Kaiga GS 3 PHWR 220 06.05.200716 Kaiga GS 4 PHWR 220 20.01.201117 Narora APS 1 PHWR 220 01.01.199118 Narora APS 2 PHWR 220 01.07.199219 Kakrapar APS 1 PHWR 220 06.05.199320 Kakrapar APS 2 PHWR 220 01.09.1995

Total 4780

3.5.5 World Scenario

There are currently 432 operable (that is, connected to electricity grids) nuclear power reactors in the world.

In the year 2010, the contribution of different types of reactors to electricity generated is as below:PWR 66%, BWR 22%, PHWR 6%, LWGR (RBMK) 3%, GCR 2%, FBR 1%

In 2012, the electricity generated through nuclear reactors as a percent of total electricity generated in different countries is indicated:France 74.8%, Belgium 51.0%, Ukraine 46.2%, USA 19.0%, Russia 17.8%, Germany 16.1.%, Pakistan 5.3%, India 3.6%,


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4 Fuel Materials

4.1 INTRODUCTIONA nuclear reactor is basically a heat source in which energy is released through the fission of an isotope of uranium or plutonium. Afissionable atom located on a crystal lattice in say, a pellet made of UO 2 crystals undergoes fission when a neutron of suitable energyis absorbed. This occurs because the resulting nucleus is very unstable and splits into two parts of roughly equal mass. For example,

235U92 +1n0

140Ba 56 +93Kr 36 + 3

1n0

235U92

+ 1n0 144Xe

54 + 90Sr

38 + 2 1n

0

235U92 +1n0

144Cs 55 +90Rb 37 + 2

1n0

There is a discrepancy in mass between the two sides of this equation, which corresponds, to an energy release of about 200 MeV.Most of this energy is imparted to the two fission fragments, which leave the site of fission very rapidly, traveling in straight lines oflength about 10 microns before coming to rest. In doing so they impart their energy to the parent lattice, essentially as thermalvibrations, and incidentally causing considerable damage to the lattice. It is this energy, plus a smaller amount arising from othersources, which represents the heat source in the nuclear fuel which must be converted to a more useful form.The design of the reactor core involves reactor physics, engineering (heat transfer, fluid flow and stress analysis), materials science,safety and economic inputs. The finished product represents a compromise or optimization between these many different factors. Thecore and hence the fuel element are designed, built, operated and refined in an iterative process.The environment in the reactor core varies from point to point. The fission rate peaks at the core centre and drops off all around

because of neutron leakage. Generally the coolant flows upward through the core so that there is a temperature rise from the bottom tothe top of the core. Since the phenomena in the fuel and cladding materials are sensitive to temperature and to fission rate or neutronflux, each point in the core will behave differently from others. This makes fuel element design and development particularly difficult.The fuel element is the fundamental building block of the reactor core. It contains a discrete quantity of nuclear fuel. The mostcommon geometry for these fuel elements is a long cylindrical rod, although plates are some times used, e.g., in research reactors. Thecoolant flows under forced convection over the cladding to remove the heat (cool the fuel element) and transports it to a heatexchanger. A research reactor has to generate a high flux, but need not generate high coolant temperatures. The high specific powerand surface heat fluxes are best handled by a plate geometry with a high surface are to volume ratio and by an aluminium-based fuel,which has good neutronic properties.The reactor core is made up of an assembly of fuel elements. Early designs used individual fuel rods, but is now standard practice togroup many rods together into a bundle or subassembly which is usually enclosed in a metallic box or duct that acts as a flowseparator and as a structural member. A subassembly is a convenient sized unit for moving fuel in and out of the reactor.

4.2 Fuel Assembly Design

The general objectives in the design and engineering of reactor fuel assemblies are to provide: A geometric arrangement of fissile, fertile and other materials that can sustain a nuclear chain reaction over a period of

several years Adequate heat transfer and fluid flow characteristics Failure free fuel pins that will contain radioactive products over the desired burnup life time, through normal and expected

transient operations, and under postulated accident conditions Economy that will help nuclear power be competitive with other energy sources Fabrication processes for efficient production of standardized, quality controlled units

Typical (interactive) limitations on fuel assemblies include:Temperatures prevent melting, excessive component expansion, fission product release, and damaging chemical reactions amongfuel, clad, and / or coolant.

Cladding stress and strain balance coolant pressure, differential pellet clad expansion, pellet swelling, fission gas pressure, and provision for neutron poisons.

The earliest reactor fuel elements were made of uranium metal. However they tended to suffer from unacceptably large dimensionalchanges as a result of thermal cycling radiation damage. Alloying elements capable of limiting the dimensional instability are oftenfound to cause excessive parasitic neutron capture, especially for thermal neutrons.The ceramic uranium and plutonium dioxide fuels have emerged as favourites for commercial reactor systems. They have satisfactoryradiation damage and fission product retention properties upto high burn up levels. They are essentially inert to high temperaturereactor coolants and have little poisoning effect in the core. Carbide and nitride fuels have received attention for similar reasons.


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4.3 Comparision of properties of different fuel materials

U UC UN U 3Si UO 2 Th. Density, g/cc 18.95 13.63 14.32 15.58 10.96U density, g/cc 18.95 12.97 13.52 15.0 9.67Therm. Cond at 500 C, W.cm/C 0.27 0.23 0.20 0.25 0.04Melting pt or phase change, C 1130 2330 2800

Decomposition

at 1 atm N 2

930Peritectoid

reaction

2800

4.4 Advantages and disadvantages

Advantages of UO 2 fuel High melting point No phase change upto melting point Good corrosion resistance in hot water Good chemical stability Good fission product retention Excellent irradiation stability Excellent compatibility with cladding

Disadvantages of UO 2 fuel Relativley low fissionable atom density. Low thermal conductivity. Brittleness. Low thermal shock resistance.

Advantages of Uranium metal fuel High uranium atom density results in small core volume. High thermal conductivity facilitates good heat transfer Good fabricability and machinability Higher breeding ratio in the case of fast reactor fuel (U+Pu+Zr)

Disadvantages of Uranium metal fuel

High chemical reactivity with water under operating conditions. The products of corrosion are more bulky than the metal. Dimensionally unstable due to anisotropy of alpha uranium. Inability to restrain the nucleation of gas bubbles which cause swelling.

.5 WATER REACTOR FUEL

The commercial power reactor field is dominated by three main types: PWRs, BWRs and HWRs. All these systems use UO 2 as fueland Zr 2 or 4 as cladding.4.5.1 Pressurized Water Reactor fuelIn all PWRs the fuel rods are assembled into a square array held together by spring clips or spacer grids and by nozzles at the top and

bottom. The assembly includes a number of control rod guide tubes in which approximately 16 control rods slide; these rods are

connected to a common actuator rod at the top of the assembly. This is known as rod cluster control (RCC). The PWR assembly isopen at the sides, allowing cross-flow of the coolant, independent expansion of the structural components and the fuel rods, and easyvisual inspection before and after discharge.The individual fuel rods vary in diameter from 9.63 mm to 11.8 mm and are between 3.71 m and 4.09 m in length. The active fuellength is shorter by about 230 mm, which represents the plenum length. The Zr 4 cladding thickness is about 0.6 mm. The UO 2 pelletshave a density of between 94 and 95%.

4.5.2 Boiling WaterReactor fuelThe BWR fuel assemblies differ from the PWR in a number of respects. Most importantly, the control rods are external to the fuelassemblies; they are cruciform shaped plates that move in the space between four fuel assemblies. The fuel assemblies thereforecontain only fuel rods, arranged in 8x8 arrays. These rods are spaced by zircaloy grids. There are 8 tie rods, which hold the top and the


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bottom of the assembly together. The assembly is contained within a zircaloy box or channel which confines the coolant flow, that is,there is no cross-flow. BWR fuel rods are clad in Zr 2. The rod diameter and the cladding thickness are somewhat larger than inPWRs; ~ 12.5 mm OD and 0.864 mm thickness. The fuel column length and the plenum are similar to that in PWRs.

TAPS 1,2 Fuel: 284 fuel assemblies per coreFuel rod array 6x6 Pellet dia. 12.4 mmWt. Of U/assembly 139 kgClad material Zircaloy 2, thickness 0.89 mmControl material Boron Carbide granules in SS tubes

No. of Control blades 69

Improvement in BWR Fuel:Increasing the helium pressure from 1 atm to 2.5 atm to prevent clad collapse under coolant pressure in the event of abnormal axialshrinkage of fuel due to in reactor densification. More effective drying of the pellets and loaded element to prevent hydriding. Getters(for example, reactive alloy of Zr, Ni and Ti) have been used in some countries which get hydrided sacrificially, thereby protecting thecladding.

4.5.3 Heavy Water Reactor fuelThe fuel is natural enrichment UO 2. Because the fuel is not enriched, it must be loaded and unloaded on-power. The coolant tubes arearranged horizontally to make this operation easier and the fuel assemblies are fairly short (50 cm) in order to achieve a higher mean

burn-up and because the core length is so great (~6m). The pellets are sealed into 13.1 mm diameter rods with 0.4 mm wall thickness. Note that the wall thickness is only half that of the smaller diameter LWR rods. Each rod has bearing pads welded to its OD. in anumber of places to separate it from its neighbours. 37 rods are bundled together and welded to two end plates. The fuel residencetime in the core is ~ 470 full power days to reach a design mean maximum burn-up of 7 GWd/t(U)

The CANDU assemblies are slightly less than 0.5 m in length to be consistent with the on line refueling strategy. This short lengthdoes not require that the cladding tubes be free standing like those for the LWR. The clad is thin and is allowed to creep down on tothe pellets. A combination of low burn up for the natural uranium fuel and the short length of the fuel pin limits fission gas productionand hence the need for the plenum space designed into LWR pins. The bundle structure is simple because the pressure tube supportsthe fuel bundle and all reactivity control mechanisms are external to the fuel channel.

RAPS Fuel: 12 bundles per channel, 306 channels

RAPP bundle: the six elements of inner ring and the six alternate elements; of the outer ring are wrapped with 0.05 dia zircaloy -2wire in a helical manner to promote inter-sub-channel mixing of the coolant. The other six elements of the outer ring have spacer


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wires welded on to them. In addition, all the twelve outer elements are provided with wrapped bearing pads of 0.065 dia zir caloy-2wire, which space the bundle from the coolant tube and reduce the wear on the coolant tube during the fuel charging operation.

RAPP 235 MW fuel bundle details:

I) UO 2 : No. of pellets/tube = 24+2 No. of pellets/bundle = 456Weight of pellets/tube = 24X33 = 795 = 800 g.Weight of pellets/bundle = 15.2 Kg.

II) Zircaloy components :

Tubes (19 Nos) 1079.2 gEnd plugs (38 Nos) 159.6 gEnd plates (2 Nos) 33.2 gSpacer wire .05 dia 93.0 gBearing pad wire 0.065 dia 37.0 g

III) Reactor charge

Weight of finished fuel bundle 16.7 kg. No. of bundles/channel 12 No. of channels in the reactor 306 No. of bundles in the reactor 3672 (306 X 12)Wt of UO2 in the reactor 56 tonnesTotal no of bundles required

before start up 4000Total fuel inventory 60 tons

19 element bundle 37 element bundle1 UO2 pellet diameter, mm 14.30 12.202 Zr-4 clad outer dia, mm 15.25 13.083 Length of fuel element, mm 493 4934 Wt of UO2 per bundle, kg 15.2 22.5

5 Number of spacer pads 72 156 (3types)6 Number of bearing pads 36 547 Number of welds per bundle 328 6228 Fuel bundles in initial core 306 x 12 = 3672 392 x 13 = 50969 Reload fuel bundles per anum 2400 405010 Electricity generated per bundle, kwh 640,000 926,000


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4.5.4 Typical composition of Jaduguda ore % by weight

U3O8 0.07,SiO 2 67.2FeO 6.37Fe2O3 7.87Al2O3 5.5TiO2 0.66MnO 0.13CaO 5.40MgO 2.20P2O5 1.05S 0.79

The ore from Jaduguda, Bhatin and Narwapahar Mines are processed in the centralized processing plant (Mill) located close toJaduguda Mine. Uranium is extracted from ore in the Jaduguda Mill by hydro-metallurgical process. After three stages of crushing, thecrushed ore undergoes two stages of wet grinding. The slurry thus obtained is pumped to the leaching pachucas for dissolution ofUranium. The leached slurry is filtered to obtain Uranium liquor.The Uranium liquor is purified and concentrated by ion exchange method. The Uranium is then precipitated from this concentratedliquor as magnesium Di- Uranate, generally known as YELLOW CAKE. This is thickened, washed, filtered and dried in the spraydryer and finally packed in drums and then sent to Nuclear Fuel Complex at Hyderabad for further purification and processing intoUO 2 pellets.

Typical Composition of MDU concentrate received at NFC from Jaduguda, % by weight

U3O 8 68-75Fe 0.4-0.6Si 3.0-6.5Ca+Mg 6.0 9.0PO 4

= 0.1 0.4SO 4

= 0.5 1.5

4.5.5 Steps in manufacture of metallic fuel for CIRUS

1. Preparation of UF 4 by fluorination of UO 2.2. Bomb reduction of UF 4 with Mg at 600 C to get uranium metal ingot (44 kg).3. Vacuum induction melting (removes N, H, O and F. Graphite crucible coated with alumina is used to prevent carbon pickup.

Dross and oxides float).4. Bottom pour and cast (75 mm diameter, 900 mm long billet).5. Hot rolling at 630 C with 5 to 11% reduction per pass6. Beta heat treatment (heat to 730 C and quench in water)7. Roller straightening8. Centreless turning9. Thread machining of ends10. Centreless grinding11. Degreasing, pickling, washing and drying12. Al inner plugs screwed to ends13. Al finned tube

14. Draw bench15. Sheath rolling over end-plugs16. Al outer plugs screwed to inner plugs17. End closure TIG welding

4.5.6 Steps in manufacture of ceramic fuel for Power ReactorsDissolution of magnesium diuranate cake in nitric acid to get slurry.Solvent extraction of the crude uranyl nitrate to get pure uranyl nitrate, using tributyl phosphate diluted in kerosene.Controlled precipitation of ammonium diuranate.Drying, calcination of the cake to get U 3O8.Reduction of the U 3O8 to UO 2 powder. (Frequently, the calcinations and reduction steps may be combined and called DirectReduction).


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Stabilisation (passivation) of UO 2 powder by controlled exposure to low oxygen.Compaction of cylindrical pellets, sintering and finish grinding.Loading of pellets into zirconium alloy tubes and welding. Resistance welding for the thinner (0.38 mm) PHWR cladding, TIGwelding for the thicker (0.83) BWR cladding.Welded assembly in the case of PHWR and mechanical assembly in the case of BWR.Inspection at every stage.

4.5.7 Improvements in PHWR Fuel:Single side dished unchamfered pellets have been replaced by Double dished and chamfered pellets.Scooped end plug to reduce zirconium content in the fuel assembly and to prevent heat transfer from the pellet at the end of theelement.Graphite coating on the inside of the fuel tube to minimize friction during pellet clad interaction.

4.5.8 KAMINI Metallic FuelPlate type, 20% U233, 80% Al, 8 g of U233 per plate, 8 plates per assembly, 9 assemblies per core

Aluminium was chosen because of its low parasitic thermal neutron absorption cross section, low cost, ready availability, easyfabricability, adequate mechanical strength, excellent irradiation behaviour, and corrosion resistance in water upto 100 C. Light wateracts as coolant and moderator while BeO acts as reflector. Handling of U 233 is as difficult as that of Pu 239, because of theaccompanying U 232 which has strong gamma emitting daughter products.

Steps in manufacture of KAMINI FuelMaster alloy, Melt, Cast, Roll, Picture frame, Roll bond

4.5.9 Comparison of fuel elements for thermal and fast reactors

4.5.10 MOX fuel in thermal reactorsMOX fuel is used in thermal reactors as well as fast reactors. About 30 thermal reactors in Europe (Belgium, Switzerland, Germanyand France) are using MOX and a further 20 have been licensed to do so. Most reactors use it as about one third of their core, butsome will accept up to 50% MOX assemblies. In France, EDF aims to have all its 900 MWe series of reactors running with at leastone-third MOX. Japan aims to have one third of its reactors using MOX by 2010, and has approved construction of a new reactor witha complete fuel loading of MOX.
http://en.wikipedia.org/wiki/Thermal_reactorhttp://en.wikipedia.org/wiki/Thermal_reactor


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Licensing and safety issues of using MOX fuel include: As plutonium isotopes absorb more neutrons than uranium fuels, reactor control systems may need modification. MOX fuel tends to run hotter because of lower thermal conductivity, which may be an issue in some reactor designs. Fission gas release in MOX fuel assemblies may limit the maximum burn-up time of MOX fuel.

About 30% of the Plutonium originally loaded into MOX fuel is consumed by use in a thermal reactor. If one third of the core fuelload is MOX and two-thirds uranium fuel, there is zero net gain of plutonium in the spent fuel. All plutonium isotopes are either fissile or fertile, although plutonium-242 needs to absorb 3 neutrons before becoming fissile curium-245; in thermal reactors isotopic degradation limits the plutonium recycle potential. About 1% of spent nuclear fuel from current

10sep2014 Nuclear Fuel Cycle Notes Balakrishna Palanki

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